TW202238623A - High energy nuclear fuel, fuel assembly, and refueling method - Google Patents

Abstract

Provided herein is a nuclear fuel assembly for a pressurized water reactor. The nuclear fuel assembly comprises: a plurality of nuclear fuel rods configured to contain a fissile material, wherein the nuclear fuel assembly is configured such that a hydrogen to uranium ratio for the fuel assembly, when coolant and the fissile material are present under operating conditions, is at least 4.0. Also provided herein is a method for refueling a pressurized water nuclear reactor comprising a nuclear fuel assembly of the present disclosure.

Description

High-energy nuclear fuel, fuel assembly, and method for refueling

This case is about high-energy nuclear fuel, fuel assembly, and methods of refueling.

In a nuclear reactor, free neutrons produced by the fission of a cleavable material (eg, nuclear fuel comprising ²³⁵ U and/or ²³⁹ P) help generate energy by sustaining a chain reaction of fragmentation of the cleavable material. This chain reaction produces heat, which can be used to generate energy in, for example, nuclear power plants. However, over time, the stock of frackable material undergoes fragmentation and becomes depleted, requiring refueling and/or other maintenance to maintain safe and efficient power generation. Refueling and/or other maintenance operations add to the cost of generating electricity from nuclear power plants. There is a need to minimize the materials and time required to accomplish such operations and/or minimize the frequency with which such operations are required.

This case provides a nuclear fuel assembly for a pressurized water reactor. The nuclear fuel assembly comprises a plurality of nuclear fuel rods organized to contain fissile material, wherein the nuclear fuel assembly is configured such that the fuel assembly has a hydrogen to uranium ratio when coolant and fissile material are present under operating conditions Department is greater than 4.0.

This case also provides a method for refueling a pressurized water nuclear reactor including the nuclear fuel assembly of the present disclosure. The method includes, for example, refueling the pressurized water nuclear reactor at cyclical intervals of 24 months.

A nuclear fuel assembly as disclosed herein can optimize the technical and economic aspects of fragmentation chain reactions for use in high energy, high discharge burnup applications such as 24 month fuel cycles. This optimization may allow more energy to be extracted from a given mass of crackable material while maintaining compliance with technical and permitting requirements for plant operation. In addition, the various fuel assembly configuration options of the present disclosure have considered the effects of fuel assembly configuration on neutron moderation, temperature, and safety protocols to provide improved safety and efficiency of nuclear reactor systems. Other benefits are also disclosed in the text.

It should be understood that the inventions described in this disclosure are not limited to the examples outlined in this Summary. Various other examples are described and illustrated herein.

This application is based on 35 U.S.C. § 119 (e) claiming that the United States filed the application title "HIGH ENERGY NUCLEAR FUEL, FUEL ASSEMBLY, AND REFUELING METHOD" on December 7, 2020 The benefit of Provisional Application No. 63/122,100, the entire disclosure of which is incorporated herein by reference.

Before explaining the various aspects of the invention in detail, it should be noted that the illustrative examples are not limited in application or use to the details of construction and arrangement of the components illustrated in the drawings and description. The illustrative examples can be embodied or incorporated in other aspects, variations and modifications, and can be practiced or carried out in various ways. Furthermore, unless otherwise specified, the terms and expressions used in this specification are chosen for the purpose of describing illustrative examples for the convenience of the reader and not for the purpose of limiting them. Furthermore, it will be appreciated that one or more of the aspects, expressions of aspects and/or examples described below may be combined with any one or more of the other aspects, expressions of aspects and/or examples described below.

Nuclear reactors produce neutron radiation to initiate and sustain the fission chain reactions responsible for their energy production. In many nuclear reactors, neutrons produced by chain reactions are moderated (eg, slowed down) to, for example, increase their likelihood of contributing to the chain reaction. These reactors include pressurized water reactors and related systems (PWR), which cool the fuel by removing heat from the split chain reaction and transfer this heat into the reactor area called primary water (indicating that this water will be Fuel rod contact and possible leakage of fuel rods including radioactive contamination) pressurized water. The primary water is used both as a coolant and as a neutron moderator. The primary water then transfers its heat to clean secondary water which is at a lower pressure than the primary water. The reduced pressure in the secondary water and the high temperature of the primary water cause the secondary water introduced into the steam generator to boil and form steam which, for example, is used to drive a steam turbine coupled to a large generator to generate electricity. Thus, PWRs use the fragmentation of fissile materials in nuclear fuel as a source of electrical energy.

Various configurations of the PWR nuclear fuel assembly are disclosed herein that allow for increased and more efficient extraction of energy (e.g., heat) from the nuclear fuel, which can increase the time between refueling cycles and reduce uranium costs, power outage costs, personnel , waste fuel disposal and/or other aspects of plant operation to reduce operating costs. Configurations of the present disclosure use, for example, newly developed nuclear fuels, fuel rods, and other components to produce nuclear fuel assemblies capable of increased energy extraction. Therefore, this case provides a nuclear fuel assembly, a method for refueling a pressurized water nuclear reactor, nuclear fuel rods, and nuclear fuel pellets. The nuclear fuel of the present disclosure may be referred to herein as "high energy nuclear fuel (HEF)," "high burnup fuel," "high enrichment fuel," and the like.

PWR uses a fuel assembly, which may include nuclear fuel rods, and, inter alia, may hold the fuel rods and guide thimbles (e.g., tubes housing control rods or instrumentation) in position and ensure proper exposure to reactor cooling during reactor operation other related components of the agent. The nuclear fuel assembly of the present disclosure may include a plurality of nuclear fuel rods configured to contain nuclear fuel. Nuclear fuel assemblies are available in 18 x 18, 17 x 17, 16 x 16, 15 x 15, 14 x 14 and other configurations, whether in square or triangular lattice, and various effective core heights. A 17 x 17 configuration can contain a total of 289 assemblies that can each be either a fuel rod or a guide thimble. In one example, a 17 x 17 configuration may contain 264 fuel rods and 25 guide thimbles.

Nuclear fuel assemblies can be designed to optimize fuel utilization for specific purposes. One optimization parameter for the fuel assembly used in the PWR is the hydrogen to uranium (H/U) ratio. The hydrogen to uranium ratio is equal to the number of hydrogen atoms divided by the number of uranium atoms of all isotopes at operating conditions within the reactor volume. Hydrogen atoms may be, for example, present in, or originate from, a component of water molecules of the PWR's coolant/moderate agent and uranium atoms may be present in, or originate from, nuclear fuel. For example, the H/U ratio can be calculated according to Equation 1 below:

Formula 1

Low H/U values result in higher average energies in the thermal neutron spectrum, thereby minimizing fragmentation and control values and favoring plutonium formation at the expense of initial enrichment of uranium requirements. A high H/U value produces a thermal neutron spectrum with a lower average energy, which increases the value of fragmentation and containment materials and favors the burning of plutonium thereby reducing the initial enriched uranium requirement.

It should be noted that the optimal H/U ratio depends on fuel cycle requirements. For example, high burnup, high enrichment, and long fuel cycles have the effect of increasing neutron absorption, which in turn requires increased moderator compared to current paradigms of power plant operation to optimize the efficiency of reactor operation. In addition, when the fuel cycle provides no value for the separated plutonium (no reprocessing), the optimal H/U ratio is also increased so that the fuel efficiently burns the plutonium grown as part of normal operation, rather than preferentially bred plutonium to for subsequent extraction and recycling. When the application of a new fuel bundle has changed significantly, the H/U ratio may be updated in the new fuel bundle. Such changes may occur in the fuels and fuel assemblies of the present disclosure. The following is a comparison of one example of a high energy fuel and fuel cycle (in accordance with the present disclosure) with another example of a fuel and fuel cycle:

the comparative fuel design Fuel Design of the Disclosure cycle length 12 months 24 months cycle energy 270 EFPD 700 EFPD Refilling combustible absorbent none Heavy boron and/or gadolinium oxide Filling type Out-In-In High Leakage In-In-Out Low Leakage power spike Low high Batch average unloading fuel consumption 33 GWd/tU 70 GWd/tU maximum fuel enrichment ＜ 5.0 w/o ^235U ＜20 w/o ^235U Reprocessing and Recycling basic assumption none H/U ratio 3.8 – 4.0 Target 4.3, Range 4.0 – 5.0

The H/U of the nuclear fuel assembly can be, for example, at least 4.1, at least 4.2, at least 4.3, at least 4.4, at least 4.5, at least 4.6, at least 4.7, at least 4.8, at least 4.9, or at least 5.0. The H/U of a nuclear fuel assembly may be, for example, 5.0 or less, 4.9 or less, 4.8 or less, 4.7 or less, 4.6 or less, 4.5 or less, 4.4 or less, 4.3 or less, 4.2 or less, or 4.1 or below. The H/U ratio may be in the range of 4.0 to 5.0, such as, for example, 4.0 to 4.5, 4.1 to 5.0, 4.2 to 5.0, 4.3 to 5.0, or 4.4 to 5.0. The H/U ratio can be, for example, in the range of 4.1 to 4.5, 4.1 to 4.4, 4.1 to 4.3, or 4.1 to 4.2. The H/U ratio can be, for example, 4.1, 4.2, 4.3, 4.4, or 4.5.

The H/U ratio can determine, at least in part, how much energy can be obtained from the nuclear fuel before maintenance and/or fuel replacement is required. The energy gained ("burnup") can be measured in gigawatt days per metric ton of enriched uranium (GWd/tU). For example, a higher H/U ratio may mean that more neutrons produced by fission will be moderated to a level suitable for capture by uranium nuclei. Various configurations of nuclear fuel assemblies are described herein that can increase fuel burnup and/or increase the H/U ratio of the PWR.

When contained within the fuel rods, the nuclear fuel may comprise any fissile material that is chemically compatible with the nuclear fuel rods of the nuclear fuel assembly and the coolant of the reactor. For example, nuclear fuel may contain uranium-containing ceramic crackable material. Uranium-containing ceramic crackable materials may include, for example, uranium silicide (eg, U ₃ Si ₂ , U ₃ Si ₅ , U ₃ Si); uranium nitride (eg, UN, U ¹⁵ N); uranium carbide (eg, UC ); uranium boride (eg, UB _x , UB ₂ , UB ₄ ), where X is an integer (metal borides (eg, uranium boride) may have a wide variety of metal:boron ratios); uranium phosphide (eg, , UP); uranium sulfide (eg, US ₂ ); uranium oxide (eg, UO ₂ , UCO, U ₃ O ₈ , UO ₃ ); or a mixture of any of these. Nuclear fuel may also contain mixtures of fissile or fertile elements, such as mixtures of uranium and plutonium, uranium and thorium, and mixtures of uranium, plutonium, and/or thorium with other actinides such as uranium, aurium, thorium, and the like.

The nuclear fuel may ^contain235U in any enrichment as required by the core design requirements. For example, up to 5% by weight of ²³⁵ U may be used, based on the total weight of uranium in the frackable material. The core design requirements can be met using fuel with an overall < 5.0 w/o ²³⁵ U. However, when used in combination with highly absorbent components within the fuel assembly and/or when the fuel cycle length and/or fuel dump burnup are increased as disclosed herein, the fuel can be designed to utilize up to 10 w/o ²³⁵ U or Up to 20 w/o ²³⁵ U enrichment for 24 month fuel cycle and high dump fuel consumption.

For example, the nuclear fuel may comprise at least 5 wt%235U and no greater than 20 wt% ^235U , based on the total weight of uranium in the ^frackable material. The nuclear fuel may comprise at least 6 wt % ²³⁵ U, at least 7 wt % ²³⁵ U, at least 8 wt % ²³⁵ U, at least 9 wt % ²³⁵ U, at least 10 wt % ²³⁵ U, at least 11 wt % ²³⁵ U, at least 12 wt % ²³⁵ U, at least 13 wt% ²³⁵ U, at least 14 wt% ²³⁵ U, at least 15 wt% ²³⁵ U, at least 16 wt% ²³⁵ U, at least 17 wt% ²³⁵ U, at least 18 wt% ²³⁵ U, or at least 19 wt% ²³⁵ U, all calculated by the total weight of uranium in the crackable material. Nuclear fuel may contain 19 wt% or less ²³⁵ U, 18 wt% or less ²³⁵ U, 17 wt% or less ²³⁵ U, 16 wt% or less ²³⁵ U, 15 wt% or less ²³⁵ U, 14 wt% 235 U at 13% or less, ²³⁵ U at 13% or less by weight, ²³⁵ U at 12% or less by weight, ²³⁵ U at 11% or less by weight, ²³⁵ U at 10% or less by weight, ²³⁵ at 9% or less by ^weight U, 8% by weight or less of ²³⁵ U, 7% by weight or less of ^235U , or 6% by weight or less of ^235U , all based on the total weight of uranium in the crackable material. For example, the nuclear fuel may comprise at least 5 wt % ²³⁵ U and no more than 15 wt % ²³⁵ U, at least 5 wt % ²³⁵ U and not more than 10 wt % ²³⁵ U, at least 6 wt % ²³⁵ U and not more than 20 wt % ²³⁵ U, at least 6 wt% ²³⁵ U and not more than 15 wt% ²³⁵ U, at least 6 wt% ²³⁵ U and not more than 10 wt% ²³⁵ U, or any other subrange, all based on the total weight of uranium in the frackable material.

The nuclear fuel may be present in the fuel rods as nuclear fuel pellets. Nuclear fuel pellets may contain fissile material. At least a portion of the nuclear fuel pellet may be an annular nuclear fuel pellet. These annular fuel pellets reduce fuel temperature and provide void volume within the fuel rod. The reduced fuel temperature and increased void volume when combined or separated can have the effect of lowering the gas pressure within the fuel rods, which is a critical limiting parameter when dealing with high burnup fuels due to the release of splits due to splitting Most of the product exists in the gas phase at the operating temperature or is volatile and forms a gas at the operating temperature. In addition, boron-containing combustible absorbents can be used within fuel rods, which emit helium due to absorption in boron. The ability to increase the void volume to accommodate the release of helium from the use of the combustible absorber can help achieve high burnup at acceptable fuel rod internal pressures. The use of annular fuel pellets for the entire fuel stack allows the core design to adjust the uranium loading and H/U ratio of the fuel assembly or fuel region without changing the hydraulic characteristics of the fuel assembly. This capability can be advantageous because it leverages fuel utilization efficiency but does not result in changes to fuel hydraulic characteristics. Changing fuel hydraulic characteristics can be costly and time-intensive.

It is contemplated that a substantial majority (eg, at least 65%, at least 75%, at least 85%) of the fuel rods within the fuel assembly may utilize annular fuel pellets. All nuclear fuel pellets can be annular nuclear fuel pellets, or none of the nuclear fuel pellets can be annular nuclear fuel pellets. The nuclear fuel pellets of the present disclosure may contain a void volume range of 2.5% to 15%. As an example, fuel pellets may have a void fraction of 2.5%, which corresponds to a pellet inner diameter of approximately 0.050 inches (1.25 mm) for a 17x17 fuel application. As another example, fuel pellets may have a void fraction of 15%, which corresponds to a pellet inner diameter of approximately 0.125 inches (3.15 mm) for 17x17 fuel.

Referring to FIG. 1 , a cross-section of an annular nuclear fuel pellet 100 of the present disclosure is shown. The pellet 100 can include an outer surface 108 and an inner surface 110 . The inner surface 110 may at least partially define a cavity 112 . The pellet 100 may include a cleavable material 114 surrounding a cavity 112 . Crackable material 114 may be disposed between inner surface 110 and outer surface 108 .

The pellet 100 can include an outer diameter extending from a point on the outer surface 108 through the center of the cavity 112 to an opposite point on the outer surface 108 (see line 106). The pellet 100 may include an inner diameter extending from a point on the inner surface 110 through the center of the cavity 112 to an opposite point on the inner surface 110 (see line 104). The nuclear fuel pellets of the present disclosure can optionally comprise an annular shape in which up to 15% of the total volume of pellet 100 is void space 112 . The total volume may include the volume of the rupturable material 114 and the cavity 112 . The pellets may have a total volume that includes up to 8%, 9%, or 10% void space 112 based on the total volume of pellet 100. Alternatively or additionally, the pellet may have a total volume comprising at least 4%, 5%, or 6% void space 112 based on the total volume of pellet 100 .

Nuclear fuel pellets of the present disclosure can optionally include a diameter 104 having an inner diameter 104 in the range of 0.065 inches to 0.075 inches (approximately 1.65 mm to 1.91 mm), such as, for example, in the range of 0.070 inches (approximately 1.78 mm). The inner diameter 104 is annular in shape. The exemplary fuel pellet 100 shown in FIG. 1 will not necessarily include any particular physical feature at lines 104,106. These lines only indicate the geometry of the inner and outer diameters. Figure 1 is an example of the present disclosure and other shapes including a central cavity can be used for the fuel pellets of the present disclosure.

HEF nuclear fuel rods of the present disclosure may comprise at least two morphologies - a first morphology with solid pellets in the enrichment region and a second morphology with annular pellets in the enrichment region. HEF fuel rods typically have five or more axial zones with at least two distinct degrees of enrichment. The axial blanket region has reduced enrichment to minimize neutron leakage from the top and bottom of the reactor. In certain examples, the axial blanket enrichment can be about 50% of the enriched region, with the top and bottom axial blankets having the same enrichment. However, in other examples, the top and bottom axial blankets may comprise different degrees of enrichment.

Axial blanketing typically utilizes annular fuel pellets where the internal void is about 25% of the solid pellet volume. Design options exist in nuclear fuel rods utilizing solid or annular axial blanketing. The fuel enrichment zone may be solid or may utilize annular fuel pellets in which the internal void is about 4-10% of the volume of the solid pellet. The enrichment zone fuel pellets contain enrichments ranging from 0.711 w/o ²³⁵ U for natural uranium up to 10 w/o ²³⁵ U or higher depending on the specific reactor core design and fuel management requirements. Enrichment zones exist within the HEF fuel rod above the lower axial blanket and below the top axial blanket. Overlapping the HEF enriched region may be the burnable absorbent (BA) region, which is generally shorter than the enriched region and is generally symmetrical around the center of the core. However, specific core design requirements may alter the BA length and centering. In at least one example, there are five axial zones within a HEF fuel rod that are axially blanketed from below, enriched with no BA or a cutback zone, enriched with BA, another enriched inverse zone , and finally the top axial blanket.

Referring to FIG. 2, a cross-sectional view of a nuclear fuel rod 200 of the present disclosure is shown. Nuclear fuel rod 200 may include an outer surface 208 and an inner surface 210 . The inner surface 210 may at least partially define a cavity 212 . Nuclear fuel rod 200 may include metal or metal alloy 214 surrounding cavity 212 . A metal or metal alloy 214 may be disposed between the inner surface 210 and the outer surface 208 . The fuel rods 200 of the present disclosure may comprise metals or metal alloys, including zirconium or zirconium alloys. For example, the fuel rod 200 may comprise an alloy containing zirconium and tin and/or niobium and optionally any of iron, tin, vanadium, and copper (e.g., ZIRLO®, Optimized ZIRLO ^™ , LT-ZIRLO™, and AXIOM ™ alloy, which is commercially available from Westinghouse Electric Company of Cranberry Twp, Pennsylvania, United States).

Nuclear fuel rod 200 may include an outer diameter extending from a point on outer surface 208 through the center of cavity 212 to an opposite point on outer surface 208 (see line 206 ). Nuclear fuel rod 200 may include an inner diameter extending from a point on inner surface 210 through the center of cavity 212 to an opposite point on inner surface 210 (see line 204 ). The example fuel rod 200 shown in FIG. 2 will not necessarily include any particular physical feature at the lines 204,206. These lines only indicate the geometry of the inner and outer diameters.

Alternatively, or in addition to the aforementioned optimization of the nuclear fuel pellet geometry, additional modifications may be made to the nuclear fuel rods themselves. Likewise, such modifications may increase the H/U ratio and/or result in more complete fuel burnup.

Referring to Figures 2 and 3, and alternatively, or in addition to the aforementioned optimization of the nuclear fuel pellet geometry, additional modifications may be made to the nuclear fuel rods themselves. The fuel rods and/or jackets may be designed such that the outer diameter 206, 306a to pitch ratio may be in the range of, for example, 0.720 to 0.745. For example, the outer diameter 306a to pitch ratio can be in the range of 0.725 to 0.745 or 0.730 to 0.740. For example, the ratio of outer diameter 306a to pitch can be 0.738. As used herein, "spacing" refers to the distance d from one center 320a of a fuel rod 300a to the center 320b of an adjacent fuel rod 300b within a fuel assembly.

The nuclear fuel rods 200 of the present disclosure, when used in a 17x17 fuel grid with a fuel rod spacing of 0.496 inches (~12.6 mm), can optionally include an outer diameter 206 in the range of 9.2 mm to 9.5 mm, such as, for example, 9.2 mm, 9.3 mm, 9.4 mm, or 9.5 mm outer diameter 206 . The nuclear fuel rods 200 of the present disclosure can optionally include an outer diameter 206 of 9.2 mm to 9.4 mm, such as, for example, a 9.2 mm, 9.3 mm, or 9.4 mm outer diameter 206 .

Combinations of fuel pellets of the present disclosure, together with fuel rods of the present disclosure may be advantageous. For example, when considering a 17x17 fuel assembly, an annular fuel pellet with a void volume in the range of 4 - 10% may correspond to an annular pellet in the range of 0.07 inches to 0.10 inches (1.8 mm to 2.5 mm) inner diameter, and may be used in fuel rods with outer diameters comprised within the range of 9.0 mm to 9.5 mm, such as, for example, 9.0 mm, 9.1 mm, 9.2 mm, 9.3 mm, 9.4 mm, or 9.5 mm. This configuration improves the H/U ratio when using a plurality of these fuel rods in a PWR.

The nuclear fuel rod 200 of the present disclosure may include a thickness which is the difference in length between the inner diameter 204 and the outer diameter 206 . The thickness may be at most 6%, at most 7%, or at most 8% of the outer diameter 206, 306a. The thickness may be in the range of 6% to 8% of the outer diameter 206, 306a. The thickness may be in the range of 6.5% to 8% of the outer diameter 206, 306a. The thickness may be 7% to 8% of the outer diameter 206, 306a. The thickness may be greater than 8% of the outer diameter 206, 306a. The thickness may be at least 0.0225 inches, which is the value used in current technology 17x17 fuel assemblies. The optimal thickness for a 17x17 fuel assembly appears to be at least 0.030 inches and is due to competing effects of reduced uranium loading, increased parasitic material, changes in fuel rod and fuel assembly axial growth and mechanical stiffness, among other effects driven by.

Jackets having such thicknesses have been found to improve the performance of the nuclear fuel assemblies of the present disclosure. For example, increasing the fuel rod jacket thickness increases the metal mass required to accommodate high burnup, 24 month cycle exposure requirements, which may be at or above core average power. The combination of increased sheath thickness and advanced sheath alloys reduces hydrogen concentration in the sheath, wear, hoop stress, and creep, all of which provide boundary sheath failure under normal conditions on a design basis. Additionally, increasing the jacket thickness as disclosed herein can result in reduced fuel volume, which when implemented with the fuel of the present disclosure, results in increased H/U ratios, which can improve uranium utilization and ultimately reduce costs. Fuel rods having jackets comprising thicknesses as described herein may be combined with fuel pellets of the present disclosure and may also comprise outer diameters as described herein.

The nuclear fuel assemblies of the present disclosure may include improved grid spacers. For example, high burn-up optimized spacers comprising advanced alloys such as those disclosed above can be utilized to minimize corrosion and growth. Alternatively or additionally, the grid/rod contact area can be maximized to increase the wear margin. Alternatively or additionally, the grid height may be increased to maximize grid crush strength.

The nuclear fuel assembly of the present disclosure may include a modified skeleton casing. For example, the skeleton sleeve may comprise a thickness in the range of 0.015 inches (approximately 0.38 mm) to 0.025 inches (approximately 0.635 mm), such as, say, 0.020 inches (approximately 0.51 mm). The skeleton sleeve may comprise a zirconium alloy as described herein.

Various parameters of the nuclear fuel assembly of the present disclosure have been described. It should be appreciated that all of these parameters may be adjusted individually or in any combination as described to provide a fuel assembly suitable for receiving and utilizing high burnup fuel on a 24 month cycle as described.

This case also provides a method for refueling a pressurized water nuclear reactor including the nuclear fuel assembly of the present disclosure. The method may include refueling the pressurized water nuclear reactor on a 24-month cycle. Refueling at 24-month cycle cycle intervals can reduce the number of power outages, time, and materials required to refuel a reactor compared to 12 or 18 month cycles. For example, a PWR comprising a nuclear fuel assembly of the present disclosure can be operated for 23 months and refueled (and not generating electricity) for 1 month. Thus, a cycle of 24 months can be performed. Additionally, the method may comprise achieving a fuel burnup of greater than 60 GWd/tU over a 24 month cycle, such as, for example, achieving a fuel burnup of greater than 70 GWd/tU. Achieving such fuel burnup can reduce the cost of electricity generation by reducing the uranium required for a given amount of electricity generated.

Various aspects of the subject matter described herein are illustrated in the following examples.

Example 1 - A nuclear fuel assembly for a pressurized water reactor, the nuclear fuel assembly comprising: a plurality of nuclear fuel rods configured to contain nuclear fuel, wherein the nuclear fuel assembly is configured such that when there is cooling under operating conditions For fuel and nuclear fuel, the fuel assembly has a hydrogen to uranium ratio of at least 4.0.

EXAMPLE 2 - The nuclear fuel assembly of Example 1, further comprising the nuclear fuel, wherein the nuclear fuel comprises a fissile material, and the fissile material comprises up to 20% by weight of ²³⁵ U based on the total weight of uranium in the fissile material .

Example 3 - The nuclear fuel assembly of any one of Examples 1 to 2, further comprising the nuclear fuel, wherein the nuclear fuel comprises a crackable material, and the crackable material comprises, based on the total weight of uranium in the crackable material, at least 5 wt% ²³⁵ U and no more than 20 wt% ²³⁵ U.

Example 4 - The nuclear fuel assembly of any one of Examples 1 to 3, further comprising nuclear fuel pellets comprising the frackable material, wherein the nuclear fuel pellets are positioned within the nuclear fuel rods, and wherein the At least a part of the equal nuclear fuel pellet is an annular nuclear fuel pellet.

Example 5 - The nuclear fuel assembly of any one of Examples 1 to 4, further comprising nuclear fuel pellets comprising the frackable material, wherein the nuclear fuel pellets are positioned within the nuclear fuel rods, and the The nuclear fuel pellets are all annular nuclear fuel pellets.

Example 6 - The nuclear fuel assembly of any of Examples 1 to 5, wherein the nuclear fuel rods comprise an outer diameter to pitch ratio in the range of 0.720 to 0.745.

Example 7 - The nuclear fuel assembly of any of Examples 4 to 6, wherein the annular fuel pellets have a void volume in the range of 4% to 15% of the total volume of the annular fuel pellets.

EXAMPLE 8 - The nuclear fuel assembly of any of Examples 1 to 7, wherein the fuel assembly has a hydrogen to uranium ratio of at least 4.3 when operating with coolant and fissile material.

Example 9 - A method for refueling a pressurized water nuclear reactor comprising a nuclear fuel assembly according to any one of examples 1 to 8, the method comprising: refueling the pressurized water nuclear reactor at cyclical intervals of 24 months fuel.

Example 10 - The method of Example 9, further comprising achieving a fuel burnup of greater than 60 GWd/tU.

Example 11 - The method of any of Examples 9-10, further comprising achieving a fuel burnup of greater than 70 GWd/tU.

Unless specifically stated otherwise, as is apparent from the foregoing disclosure, it should be understood that throughout the foregoing disclosure, discussions using terms such as "process," "calculate," "operate," "determine," "display," or the like are Refers to the operation and processing of a computer system or similar electronic computing device, manipulating and transforming data expressed as physical (electronic) quantities in the computer system's temporary registers and memories into similarly expressed computer system memory or temporary registers or Other such information stores, transmits or displays other data of physical quantities in the device.

One or more components may be referred to herein as "configurable to", "configurable to", "operable/operable to", "adaptable/adaptable", "capable of", "conformable to" Wait. Those skilled in the art will recognize that unless the context requires otherwise, "configured to" can generally encompass active state components and/or inactive state components and/or standby state components.

Those skilled in the art will recognize that terms used herein generally and in the appended claims in particular (e.g., the subject of the appended claims) are generally intended as "open" terms (e.g., the term "Including" should be interpreted as "including but not limited to", the term "having" should be interpreted as "at least" and the term "includes" should be interpreted as "including but not limited to" Wait). It will be further understood by those skilled in the art that if a particular number of an incorporated claim statement is intended, such an intent will be explicitly stated in the claim claim, and in its absence there is no such intent. For example, as an aid to understanding, the following appended claims may contain usage of the introductory phrases "at least one" and "one or more" to introduce the claim statement. However, use of such phrases should not be construed to imply that the introduction of a claim statement by the indefinite article "a" or "an" limits any particular claim statement containing such an introduced claim statement to Claims containing only one of these statements, even when the same claim includes the introductory phrase "one or more" or "at least one" and an indefinite article such as "a or an" (eg, "一(a and/or an)" should generally be interpreted to mean "at least one" or "one or more"); the same applies to the use of definite articles used to introduce claim claims.

Furthermore, even if a particular number of incorporated claim claims is expressly stated, those skilled in the art will recognize that such statements should generally be construed to mean at least that number stated (e.g., no modifier with no other modifiers). Statement "two statements" generally means at least two statements or two or more statements). Furthermore, in those cases where conventions like "at least one of A, B, and C, etc." are used, generally such constructions are intended so that those skilled in the art should understand the meaning of the convention (e.g., "has A, B, etc. and C" would include, but not be limited to, having only A, only B, only C, A and B together, A and C together, B and C together, and/or A, B and C together, etc. system). In those cases where conventions like "at least one of A, B, or C" are used, generally such constructions are intended so that those skilled in the art should understand the meaning of the convention (e.g., "has A, B, or C A system of at least one of "will include, but is not limited to, a system with only A, only B, only C, A and B together, A and C together, B and C together, and/or A, B and C together, etc.) . Those skilled in the art will further understand that, unless the context dictates otherwise, no matter in the description, claims or drawings, usually two or more separate words and/or phrases of alternative terms should be The possibility of including one of these terms, either of these terms, or both terms is understood to be encompassed. For example, the phrase "A or B" should generally be read to include the possibilities of "A" or "B" or "A and B."

With respect to the appended claims, those skilled in the art will appreciate that the operations recited therein can generally be performed in any order. Furthermore, although the various operational flowcharts are presented in a sequence, it should be understood that the various operations may be performed in an order different from that illustrated or may be performed concurrently. Examples of such alternate order may include overlapping, interleaving, interrupting, reordering, incrementing, preparatory, supplementary, synchronous, reverse, or other modified order, unless the context dictates otherwise. Furthermore, terms such as "in response to," "in relation to," or other past tense adjectives are generally not intended to exclude such variants, unless the context dictates otherwise.

It is worth noting that any reference to "an aspect", "an aspect", "an example", "an example" and the like means that a particular feature, structure or characteristic described in conjunction with the aspect is included in at least in one form. Thus, appearances of the phrases "in one aspect," "in an aspect," "in an example," and "in an example" throughout this specification do not necessarily all refer to the same aspect. Furthermore, the particular features, structures or characteristics may be combined in any suitable manner in one or more aspects.

Any patent application, patent, non-patent publication or other disclosure material referenced in this specification and/or listed in any Application Data Sheet (Application Data Sheet) is hereby incorporated by reference at To the extent the incorporated material is not inconsistent with this specification. Accordingly, and to the extent necessary, the disclosure as expressly stated herein supersedes any contradictory material incorporated herein by reference. Any material, or portion thereof, that is purportedly incorporated herein by reference but that contradicts existing definitions, statements, or other disclosure material set forth herein will only be distinguished between the material it incorporates and the existing disclosure material. Incorporate to the extent contradictory.

The terms "comprise" (and any form of comprising, such as "comprises" and "comprising"), "have" (and any form of having, such as "has" and "having"), "include" (and any form of including, such as "includes" and "including"), and "contain" (and any form of , such as "contains" and "containing (containing)") are open linking verbs. Thus, a system that "comprises", "has", "includes" or "contains" one or more elements possesses those one or more elements, but is not limited to possessing only one or more of those elements. Likewise, an element of a system, device or appliance that "constitutes", "has", "includes" or "includes" one or more features possesses such one or more features, but is not limited to only possessing such one or more features. multiple features.

Unless specifically stated otherwise, the terms "about" or "approximately" as used herein mean an acceptable error for a particular value as determined by ordinary skill in the art, the error depending in part on the manner in which the value was measured or determined . In certain embodiments, the term "about" or "approximately" means within 1, 2, 3, or 4 standard deviations. In certain embodiments, the term "about" or "approximately" means 50%, 20%, 15%, 10%, 9%, 8%, 7%, 6%, 5% of a given value or range. %, 4%, 3%, 2%, 1%, 0.5% or 0.05%.

Any recitation of a numerical range throughout this specification is intended to include all subranges subsumed therein. For example, the range "1 to 10" is intended to include all subranges between (and including) the recited minimum value of 1 and the recited maximum value of 10, i.e., have a minimum value equal to or greater than 1, and has a maximum value equal to or less than 10.

In summary, numerous benefits have been described that result from the adoption of the concepts described in this specification. The foregoing description in one or more forms has been presented for purposes of illustration and description. It is not intended to be exhaustive or to be limited to the precise forms disclosed. Modifications or variations are possible in light of the above teachings. The form or forms were chosen and described to illustrate principles and practical application, enabling one skilled in the art to utilize various forms and with various modifications as are suited to the particular use covered. The scope of claims hereby filed is intended to define the entire category.

100: Annular nuclear fuel pellet 104: line 106: line 108: Outer surface 110: inner surface 112: cavity 114: Crackable material 200: Nuclear Fuel Rods 204: line; inner diameter 206: wire; outer diameter 208: outer surface 210: inner surface 212: cavity 214: Metal or metal alloy 300a: fuel rod 300b: fuel rod 306a: outer diameter 320a: center 320b: center d: distance

Various features of the various embodiments described in this specification are set forth with particularity in the appended claims that follow. However, various specific examples of the organization and method of operation and their advantages can be understood from the following embodiments in conjunction with the accompanying drawings:

Figure 1 is a cross-sectional view of a nuclear fuel pellet according to the present disclosure.

Figure 2 is a cross-sectional view of a nuclear fuel rod according to the present disclosure.

Figure 3 is a cross-sectional view of two nuclear fuel rods according to the present disclosure.

Corresponding reference numerals indicate corresponding parts throughout the several views. The exemplifications set forth herein illustrate various embodiments of the invention in one form, and such exemplifications are not to be construed as limiting the scope of the invention in any way.

100: Annular nuclear fuel pellet

104: line

106: line

108: Outer surface

110: inner surface

112: cavity

114: Crackable material

Claims (11)

A nuclear fuel assembly for a pressurized water reactor, the nuclear fuel assembly comprising: a plurality of nuclear fuel rods assembled to contain nuclear fuel, Wherein the nuclear fuel assembly is configured such that the fuel assembly has a hydrogen to uranium ratio of at least 4.0 when coolant and nuclear fuel are present under operating conditions. The nuclear fuel assembly according to claim 1, which further comprises the nuclear fuel, wherein the nuclear fuel comprises a crackable material, and the crackable material comprises at most 20% by weight of ²³⁵ U based on the total weight of uranium in the crackable material. The nuclear fuel assembly according to claim 1, which further comprises the nuclear fuel, wherein the nuclear fuel contains a crackable material, and the crackable material contains at least 5% by weight ²³⁵ U and not greater than the total weight of uranium in the crackable material 20% by weight ²³⁵ U. The nuclear fuel assembly of any one of claims 1 to 3, further comprising nuclear fuel pellets, the nuclear fuel pellets comprising the fissile material, wherein the nuclear fuel pellets are positioned within the nuclear fuel rods, and wherein the nuclear fuel At least a portion of the pellet is an annular nuclear fuel pellet. The nuclear fuel assembly of any one of claims 1 to 4, further comprising nuclear fuel pellets, the nuclear fuel pellets comprising the fissile material, wherein the nuclear fuel pellets are positioned within the nuclear fuel rods, and wherein the nuclear fuel The pellets are all annular nuclear fuel pellets. The nuclear fuel assembly according to any one of claims 1 to 5, wherein the nuclear fuel rods comprise an outer diameter to pitch ratio in the range of 0.720 to 0.745. The nuclear fuel assembly of any one of claims 4 to 6, wherein the annular fuel pellets have a void volume in the range of 4% to 15% of the total volume of the annular fuel pellets. 7. The nuclear fuel assembly of any one of claims 1 to 7, wherein the fuel assembly has a hydrogen to uranium ratio of at least 4.3 when operated with coolant and fissile material. A method for refueling a pressurized water nuclear reactor comprising a nuclear fuel assembly according to any one of claims 1 to 8, the method comprising: The pressurized water nuclear reactor is refueled at cyclical intervals of 24 months. The method of claim 9, further comprising achieving a fuel burnup greater than 60 GWd/tU. The method of claim 9 or claim 10, further comprising achieving a fuel burnup greater than 70 GWd/tU.

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